Procedures were developed for fabricating highperformance radiators and heat exchangers for the Aircraft Nuclear Propulsion (ANP) Program. These components, which contain multitudes of tube-to-tube sheet and tube-to-fin joints, are similar in design to those under consideration for a variety of space vehicle applications. In order to ensure reliability of the tube-to-tube sheet joints, techniques producing welds of extremely high quality were used and back brazing of the welds with a suitable alloy was incorporated. High-temperature brazing was also incorporated to attach high-conductivity fins to Inconel tubes in the radiators. The selection of a suitable brazing alloy for these applications was dependent upon several factors, including corrosion and oxidation resistance, flow point, and mechanical properties. A Ni- Si-B alloy was found to be adequate from all these considerations. Special brazing procedures were developed to obtain satisfactory flowability of this brazing alloy on tube-to-fin joints. The suitability of these fabrication procedures for the very stringent service conditions to which the radiators and heat exchangers were subjected was demonstrated by testing full-size components under operating conditions. (auth)

Organization of the ORNL Operations division for administration of the Oak Ridge Research Reactor, the Low Intensity Testing Reactor, and the Oak Ridge Graphite Reactor is described. (J.R.D.)

Criteria and guide lines are presented for the design, construction, and operation of safe, economical, and efficient glove boxes and associated facilities based upon present conditions and anticipated changes. Comprehensive discussion of glove box materials and components, safety and fire prevention methods, health physics problems, operational considerations, and brief descriptions of AEC installations are included. (39 references) (C.H.)

The fast neutron fluxes in selected positions of the MTR were measured for Cycle 146. The measurements were made at the beginning, throughout, and at the end of the cycle (564 Mwd). Vertical traverses for each position monitors are shown. (auth)

Studies are described of UO/sub 2/ powders to determine the relation between properties of the powders and sinterability. The main effort during the period was devoted to correlation of the powder particle size distribution with the sintered density of pellets. The pressure applied during the green forming step of pellet production is shown to have a varied and considerable effect on ceramic particle size distribution in the pellets. (J.R.D.)

The Oak Ridge National Laboratory has three generalpurpose research reactors which accommodate testing loops, target irradiations, and beam-type experiments. Since the experiments must share common or similar facilities and utilities, be designed and fabricated by the same groups, and meet the safe safety criteria, certain standards for these have been developed. These standards deal only with those properties from which safety and economy of time and money can be maximized and do not relate to the intent of the experiment or quality of the data obtained. The necessity for, and the limitations of, the standards are discussed; and a compilation of general standards is included. (auth)

Pilot plant studies were conducted to develop a continuous process for recovering uranium from nichrome fuels (HTRE). The process consisted of dissolution of fuel in mixed HCl--HNO/sub 3/ solution, removal of the chloride ion by stripping with HNO/sub 3/ in a packed column, and then recovery of the uranium by TBP solvent extraction. Recovery of uranium from nichrome fuels at satisfactory rates and efficiencies can be obtained by this process. Reactor fuels containing large amounts of silica may present a solids problem during chloride removal. Titanium is a suitable material of construction for the dissolver, stripping column, and their associated equipment (feed, off-gas, and product vessels). Stainless steel equipment is suitable for the solvent extraction system. (auth)

Analyses of clam shells for Sr, Sr/sup 90/, and Ca are reported. The data include 208 Sr, 80 Sr/sup 90/, and 35 Ca analyses. Information on age of the clam and shell weight are also included because the Sr concentration in some shells is affected by age and growth rate. A detailed description of sample treatment and preparation is also included. (auth)

A two-foot-square fluidized bed calciner was designed and operated to convert aqueous, highly radioactive wastes into granular solids. The calciner exceeded its designed feed capacity of calcining, at a bed temperature of 400/sup 0/C, 100 liters per hour of aluminum nitrate solution simulating wastes from the reprocessing of spent aluminum-uranium alloy reactor fuel. Heat was supplied to the calciner by circulating NaK with an electromagnetic pump, at temperatures up to 1400/sup 0/F, through a tubular heat exchanger placed directly in the fluidized bed of solids. The results of ten runs are presented and discussed. Equipment development progressed to the point where a continuous, trouble-free, operating period of 43 days was attained. Several properties of the alumina product were routinely measured, and some of the effects of the calciner operating variables on these properties were determined. The production of both amorphous and alpha crystalline material was found to be possible; the crystalline nature of the product had a profound effect on product properties and off-gas loading.

This report, for July 1962 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; and safety and security.

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This report, for March 1962 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; weapons manufacturing operation; and safety and security.

The test detailed in this report authorizes the installation of sampling tubes in the vacant-positions in one of the two front-to-rear graphite stringers to be installed in KW Reactor in January 1963. These lines will permit withdrawal of a gas sample from the active zone at a point where the graphite temperature is being measured. Analysis of this gas sample on the gas chromatograph and on a new catalytic oxygen-hydrogen instrument is expected to yield valuable information on graphite burnout versus temperature and gas quality.

The purpose of this document is to present the numerical results of a production test of the last ditch water backup system performed at B and C reactors on December 15, 1961. The main purpose of the production test was to determine more accurately the capacity of the export system for supplying flow to a dual reactor area under several simulated emergency conditions where BPA power to all areas has been interrupted and the steam supply to one old area has been lost. In addition, the flow capacity of the high tanks to,the reactor was tested alone and in parallel with the export system, and the action of the surge suppressors at all areas was recorded when cycling of the surge suppressors was deliberately induced.

A new outlet elbow fitting has been designed for use on C reactor to replace those elbows which fail because of stripped threads where the elbow connects to the rear nozzle barrel. There are two models of this threadless elbow, one providing for a water thermocouple only, and one providing for a thermocouple and an RTD. Flow tests have been conducted to determine the pressure drop across these outlets and the pressure drop has been compared to the pressure drop across the previously used outlet elbow and pigtail assembly. Results of these tests and comparisons are shown.

The objective of this production test was to evaluate the irradiation behavior of N. Reactor fuel elements at conditions equivalent to or more severe than those expected in N Reactor. The thermocouple train consisted of eleven 23-inch NAE1 elements (replicas of an N Reactor fuel element) plus five 16-inch tubular perfs. The elements were enriched to 0.947 per cent. They were coextruded Zircaloy-2 jacketed self-supported uranium tubes concentrically nested one within the other to provide an element with four heat transfer surfaces and three coolant channels. For specific dimensions, refer to Production Test IP-477-A.

This report, for June 1962 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; and safety and security.

The SNAPSHOT Safety Program, as described in this document, is therefore formulated to align safety studies performed by separate agencies and their contractors. At the present stage in program development, the plan is principally concerned with the initial SNAPSHOT flight (SNAP 10A). SNAPSHOT represents a special flight mode for the initial reactors operated in space. The flights are designed to permit early space reactor tests prior to resolving all the safety proglems that may be associated with more advanced programs. Of primary significance is that reactor operation will not be initiated until an orbit of satisfactory lifetime has been obtrained, thus ensuring the long-term decay of the fission products producted prior to re-entry.

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A superposition model is developed to predict the critical heat flux in forced convection flow. The model is applied to available experimental results in boiling water flows and good agreement is obtained between the model and test data over the multitude of geometries, flow rates, pressures, and fluid enthalpies tested to date. (auth)

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